Pressurized-water nuclear reactors incorporate a primary circuit, in which the pressurized water for cooling the reactor circulates and transfers the heat of the core to the steam generators. The primary circuit usually comprises three or four loops, on each of which is arranged a steam generator which receives, on the one hand, the pressurized water and, on the other hand, the feed water which it heats and vapourizes as a result of heat exchange with the pressurized water.
In the event of an accident in the nuclear reactor, it is immediately shut-down by introducing all the control bars of the reactor into the core in their position of maximum insertion. The reactor then has to be cooled, and this can be achieved by means of the steam generators. However, the normal steam utilization and feed water recirculation circuit, which incorporates several components such as heating devices and pumps, and which therefore has a complex structure, cannot be used for the emergency cooling of the reactor after an accident. To ensure this emergency cooling, an auxiliary circuit feeding secondary water to the steam generator is therefore put into operation for each of the loops of the primary circuit incorporating a steam generator, this auxiliary circuit then taking the place of the normal steam utilization and secondary-water recirculation circuit.
So that the nuclear reactor can be cooled under all circumstances, in particular even if a source of electrical energy is no longer available on the reactor site, cooling devices comprising only passive elements have been proposed. In particular, an auxiliary feed circuit for the emergency cooling of the reactor has been proposed, and this comprises a steam condenser located at a higher level than the relatively low level of the water of the steam generator, i.e., the equivalent overall level in the form of a liquid phase, and pipelines which respectively connect the steam outlet of the generator to the inlet of the condenser and the outlet of the condenser to the feed water inlet in the steam generator.
In this auxiliary circuit, the condenser is submerged in a tank filled with water and communicating with the atmosphere, and it is cooled as a result of the boiling of the water contained in this tank, the steam generated being discharged to the atmosphere.
Such a device is therefore capable, in principle, of cooling the reactor without the need for an external energy source, the water recovered at the condenser outlet returning to the steam generator under the effect of gravity.
However, during the cooling of the primary circuit of the reactor, the operating conditions of the steam generator and the flow of steam generated vary greatly, with the result that the condenser has to operate under essentially variable conditions, but this greatly complicates the design of this condenser which, during certain cooling phases, has in any case to operate under conditions very different from its optimum operating conditions.
On the other hand, it is possible to regulate the operating conditions of the auxiliary circuit only in so far as this does not require the use of an energy source or the involvement of facilities outside the auxiliary cooling circuit.
Finally, the mass of water contained in the cooling tank of the condensers already proposed is very large, and this greatly complicates the design of the buildings in which these condensers are installed and which in any case must be capable of withstanding earthquakes.
The need to provide an intrinsically safe emergency cooling system to prevent the extremely serious consequences of operating a reactor under conditions prevailing in the event of an accident has arisen in recent years. Such an intrinsically safe system must operate without the need for an external energy source, without an operator being involved after it has been put into service, and without a regulating working fluid; it must also have available a cooling source of virtually unlimited capacity.